IR 05000219/1988007
| ML20151J450 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/14/1988 |
| From: | Lange D, Lumb T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20151J446 | List: |
| References | |
| 50-219-88-07OL, 50-219-88-7OL, NUDOCS 8808020257 | |
| Download: ML20151J450 (71) | |
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'J.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.
38-07 (OL)
FACILITY DOCKET NO.
50-219 FACILITY LICENSE NO.
OPR-16 LICENSEE:
Gnu Nuclear Corporation P. O. Box 383 Farked River, New Jersey 08371 FACILITY:
Dyster Creek Nuclear Generating Station EXAMINATION DATES:
April 11 - 14, 1968 D um. s
[48% _ _
y/f/W CHIEF EXAMINER:
b, Sq#ior Operations Engineer Date APPROVED BY:
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__ection
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_2-/ffR David J. Lange, hief, Ta~te Operations Branch, Division of Reactor Safety
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SUMMARY: Written examinations and operating tests wers administered to one (1) senior reactor operator (SRO) candidate and three (3)
reactor operator (RO) candidates. All of the candidates passed the examinations.
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8600020257 880715
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OETAILS TYPE OF EXAMINATIONS:
Replacement EXAMINATION RESULTS:
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R0 SR0
Pass / Fail Pass / Fail
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1.
CHIEF EXAMINER AT SITE:
T. Lumb, Senior Operations Engineer
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OTHER EXAMINERS:
M. Daniels, Examiner (Sonalysts)
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S. Pu11ani, Senior Operations Engineer l
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(Eraminer in Training)
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T. Fish, Operations Engineer (Examiner in l
Training)
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3.
The following is a summary of generic strengths and deficiencies noted on the operating tests.
This information is being provided to
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aid the licensee in upgrading license and requalification training
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programs. No licensee response is required.
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l STRENGTHS l
a.
Security awareness
b.
Radiological Control procedures I
c.
Knowledge of remote shutdown equipment and procedures
DEFICIENCIES a
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Understanding of diesel generator componant operations
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b.
Reactor Operator ability to locate items in Technical
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j Specifications
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4.
The following is a summary of generic strengths and deficiencies i
noted from the grading of the Reactor Operator written examinations.
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(No generic information is available for the Senior Reactor Operator i
written examination due to the single candidate.) This information
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is being provided to aid the licensee in upgrading license and
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requalification training programs. No licensee response is
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STRENGTHS l
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a.
Knowledge of precautions for operating centrifugal pumps -
Quastion 1.06
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b.
Knowledge of thermal limit failure mechanisms, limiting conditions and limiting parameters - Question 1.08
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c.
Understanding of the effects of increasing power on core flow l
and recire pump NPSH - Question 1.09 I
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d.
Understanding of overall plant system design - Section 2 l
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i e.
Understanding of Core Spray system initiation - Question 3.05
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Knowledge of startup procedure requirements affecting control
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rod movement - Question 4.03
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g.
Understanding of the procedure for RBCCW failure response -
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Question 4.04
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Knowledge of the verification requirements for operation of j
Standby Liquid Control - Question 4.05
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Understanding of Shutdown Cooling system operational precautions
- Question 4.06
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j. Understanding of recirculation pump operational precautions -
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Question 4.09
i DEFICIENCIES
a.
Unde standing of xenon effects on reactor power following a
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transient - Question 1.01
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b.
Understanding of plant response to a rod withdrawal -
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I Question 1.10
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Understandina l
j Question 3.0$ofContainmentSpraysysteminitiation-j
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d.
Understanding of ADS operation - Question 3.09
I Knowledge of the bases behind various main turbine trip signals i
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d l
- Question 3.10
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Knowledge of 10CFR20 and Oyster Creek administrative radiation
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exposure limits - Question 4.10
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Personnt) Present at Exit Interview:
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t NRC Personnel
J T. Lumb, Senior Operations Engineer
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5. Pullant, Senior Operations Engineer
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i T. Fish, Operations Engineer l
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M. Daniels, Excminer (Sonalysts)
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l E. Collins, Resident Inspector
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Facility Personnel
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t P. Fiedler, Vice President and Directer. Oyster Creek
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E. Fitapatrick, GPU Nuclear i
J. Sullivan, Director, Plant Operations
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H. J. Lapp, Jr., Manager, Plant Training l
R. Davidson, Manager, Operator Training H. Teltt, Superviscr. License /Non-License Operator Training l
M. Heller, Oyster Creek Licensing Engineer j
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Sur. mary of NRC Coments Made at Exit Interview:
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During the administration of the written examination there was j
i distraction from technicians working in the building.
This was l
l resolved following discussion with the technicians.
The coments on j
the written examinations should be sent to the NRC within five
j working days.
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There were some problems with access to vital areas, but the l
l pr3blems were resolved promptly during the operating examiantions, l
i Health Physics and Operations personnel were cooperative.
The
j generic strengths or weaknesses noted on the operating examinations
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i were presented (see section 3 of this report).
The results of the i
j examinations would not be discussed at the exit reeting but would be j
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contained in the Examination Report.
Every effort would be made to
send the candidate's results in approximately 30 working days.
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The examiners noted a problem wit,h utiliaing the Erergency Operating
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j Procedure (EOP) flowcharts in their present form.
The flowcharts are stored rolled up and the candidates had difficulty reading them l
in the unrolled position, i
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Summary of Facility Cements Made at Exit Interview:
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The factitty training personnel comented that the operating
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exaninations were lengthy and t, hat there were fewer questions en the l
written examinations than on past examinations.
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Attacheents:
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Reactor Operator Written Examination and Answer Key
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2.
Senior Reactor Operator Written Examination and Answer Key
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Facility Comments on Written Examination after Factitty Review
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NRC Response to Facility Cor ents I
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REACTOR OPERATOR IIAM e
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HRC Ouestion. Answer and Reference
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Question 1.02 While the plant is operating at 100% power, a malfunction occurs in a Recirculation Pump MG Set Speed Cantroller, causing it to attempt to attain full speed at maximum acceleration.
DESC2IBE ANY CHANCES THAT TAKE PLACE in the PERCENT VOIDS, REACTOR POWER and
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MODERATOR TEMPERATURE in the twent*/ (20) seconds following L;se malfunction.
EXPLAIN the reasons for the changes.
J Answer 1."'
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a.
The psrcent volds decreases due to the increased recirculation flow roving the void boundary higher in the core.
b.
Reactor power rapidly incrosses to the scram setpoint due to the increased modetation in the volume where the voids had been displaced.
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Moderator temperature will ine rease due to less sub-cooling for the c.
higher recirculation flow.
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Ref<rence 1.
Oyster Creek Nuclear Genersting Station 2.
Lesson Plan 201, BNR Operating Characteristics, p. 42.
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Learning Objective 823.03 1-C-12 Epellity Cemment
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Moderator Temperature may be seen to decrease due to increksed feed flow leading to increased sub-cooling.
Should riso accept an answer of mederato-/
temperature remaining unchanged due to avexaging effects of transient and the fact that once voida are formed, the saturated system effect meintains moderator temperature relatively conv* ant.
Reference F9ne 4145J 1.0
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33C.Duestion. Amswer and Lafaranca
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Question 1.04 For each of the fc13.owir.g cop 11tions, EXPLAIN WHICH coafficient of reactivity acts FIRST to change "-
over.
Indicate if power INCREASEF or DECREASES due to the coefficient.
...sume no RPS action).
a.
The MSIV's close at 100% power.
b.
Feedwater injection temperature decreases 30 F in three (3) minutes.
s c.
A fully inserted rod drops out of the coth at power.
Answer 1.04 a.
Void coefficient power increases, b.
Moderator tamperature coefficient power decreasse.
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c.
Doppler (rael temperature) coefficient power decreases.
Reference
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Cyster Creek Nuclear Generating Station 2.
Lesson Plan 300.08, Reactivity Coefficients and Control, Rod North, pp. 17 and 28.
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Learning Objective 842.08 l
Facility commsat
M11 spelling in answar of Part b.
A decreasing moderator temperature will
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cause a power increase, not decrease.
Examiner agreed.
f.P100.08,pp.32-4h References i
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MRC Ouestion. Answer and mafarence
Question 1.05 l
Steam enters the turbine control valves at 935 psig with a st9am quality of 88%.
DETERMINE THE FOLLOWING e.
The steam temperature.
b.
The specific eethalpy of the steam.
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Answer 1.05 a.
From the Steam Tables 935 psig = 950 psia (0.25) = 538.4 r b.
hf = 534.7, hfg = 550.4, hg = 1192.9 hm = hf +.88 hfg hm = 534.7 +.88 (650.4)
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hm = 534.7 + 572.4 hm = 1107.1 btu /lb
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Referenco 1.
General E)1ctric BWR Acadsmics Series 1985 2.
Heat Transfer and Fluid Flow., pp. 5-56 and 5-57 3.
Learning Objective 852.06 1-C Faellity. comment
, S h-accept reasonable valuss using the Mollier Diagram also.
Additionally,
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shou.
consider reducing the point va*ue.
Reference Mollier Diagram 4145J 3.0
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MRC Ouestion. Answer and Reference
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Question 1.06 Running a pump at full discharge pressure and no flow is refeer4d 6,o de "dead heading".
Answer each of the questions listed below concarning the operatica of peps.
A centrifugal pump should not be run for extended periods of time a.
"dead headed". WHAT IS THE REASON for this precaution?.
J b.
WHAT design feature of the Core Spray pumps is used to minimise the possibility of "dead heading" the pump?
WHAT design feature is used on Standby Liquid Control pump to c.
minimise "dead heading" the pump?
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When starting a large centrifugal pump, the downstream system should be Illied and the discharge valve closed. WHAT ARE t'wo (2) reasons
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for this precaution?.
Answer 1.06
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If the pump is run for extended periods of time at shutoff head, a.
sufficient heat wod1d be added to the fluid to :suse cavitation and resulting internal pump damage.
c.1 A recirculation valve is installed between the pump and the d!scharge valve to cpen un low system flow.
2 A relief valve is installed between the pump and the discharge
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valve, set to lift below t h shutoff head of the pump.
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MRC Ouestion. Answer and Reference l
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Answer 1.06 (Continued)
c.1 Mith the discharge valve open, a latga mass of fluid would be moved causing a high torque and excessive starting currents, which could danaga the motor windings.
2 If the downstream system were not filled (not providing back i
pressure) the pump could go to "runout" conditions resulting in j
damage to the motor windings.
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l Reference
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1.
General Electric BWR Academics Series 1985 2.
Heat Transfer and Fluid Flow., pp. 6-108 and 6-109 3.
Learning Objective 417.02 1-A-1, 1-A-6, 1-A-8 and 1-J Facility Comment
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e In Part d of this question, another problem with starting a pump with the downstrean i
- J ng not filled is the damage that can be done by water hammer.
Should consider water hammer as an acceptable answer.
- Answer Key is misnumbered also.
Answer b.2 is in reality c, and c.1, c.2 are d.1, d.2 References None
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MRC Ouestion. Answer and Reference
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Question 1.07 The Oyster Creek Main Condenser maintains a sub-cooled condition for condensate depression.
WHAT is one (1) advantage and one (1) disadvantage of having a.
condensate depression in the condenser.
b.
For a condensate depression of 10 F, what would be the TEMPERATURE j
at Condensate Pump suction with a condenser vacuum of 27" Hg7 Answer 1.07 a.1 Assure not positive suction head to the condensate pumps.
2 Reduces plant overall efficiency.
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b.
27" Hg = 1.47 psia
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Sat temperature for 1.47 psia = 109 F 10 condensate depression = 109-10 = 99 F (+/- 1 F)
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Reference 1.
General Electric BWR Academics Series 1985 2.
Heat Transfer and Fluid Flow., p. 7-45 3.
Locraing Objective 852.06 1-G l
fac1Jity Comment
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,Should considor use of values in the temperature portion of the Steam Tabloo in addition to the pressure portion.
1.438 = 114 F and 1.51f = 116 F.
Should accept answer based on these numbers.
Additionally, should consider reducing the point value of the calculation.
Reference Table la Saturated Steams Temperature Table 4145J 6.0
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MRC Ouestion, Answer _A,nd Rafarence t
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Question 1.08 On the attached chart (Figure 2), IDENTIFY the areas indicated by letters A through E.
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Answer 1.08 a.
Fuel cladding cracking due high stress.
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b.
MLHGR c.
Clad temperature of 22 F d.
Fuel Cladding Cracking due to lack of cooling e.
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Reference
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Geocral Electric BNR Academics Seriws 1985 2.
Heat Transfer and Fluid Flow., p. 9-15
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3.
Learning Objective 852.09 1-K and 1-0 Facility Copewat In Part b of question LNGR wan misspelled MLNGR.
Reference: GE KTFF, Chapter 9, p. 9-69 I
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NRC-Question, Answer and Rafarance I
Question 2.01 An automatic initiation signal for the Isolation Condenser System has been received (High Drywell pressure or Low RPV level), and the condensers are in operation, when a steam line break occurs BETWEEN the pressure vessel and the differential pressure detectors for the "A" condenser.
a.
MILL the "A" Isolation Condenser isolate? EXPLAIN.
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b.
Da en isolation signal, WILL a LOSS of 480 volt AC power PREVENT the ISOLATION of the condenser? EXPLAIN.
c.
WILL a LOSS of Instrument and Service Air PREVENT AUTOMATIC MAKEUP to the Isolation Condenser? EXPLAIN.
Answer 2.01
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a.
Yes. The differential pressure detectors respond to a differential pressure (with steam flow passing in either direction).
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b.
No.
Each steam supply and condensate retura line has two (2) MOV isolation valves in series. One valve is 480 volts AC and the other is 125 volts DC.
On losa of 480 volt AC power the DC valves will still close.
c.
No.
An air accumulator provides air to operate the makeup valves at least six (6) times with the complete loss of Instrument and Service Air.
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Reference 1.
Dyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 23; Isolation Condenser System, j
pp. 9, 10 and 13
3.
Leerning Objective 823.23 I
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HC Question. Answer sna Rafarance
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2.01 (Continued)
Fac111ty_ Comment There is no automatic makeup to the isolation condensors.
(Part c).
OC has remotely opereted makeup valves instead of automatic. The accumulators will allow continued operation from the remote location up to six times.
Reference OPM Module 23, p. 23-17
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NRC Ouestion. Answer and Reference
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Question 2.02 During the plant startup the Main Steam Isolation Valves (MSIV's) are being opened, s.
125 volt DC power is lost to kSIV-04A Air Supply Solenoid #1.
WILL
this loss of power prevent the MSIV from being opened? EXPLAIN.
b.
On a loss of Instrument and Service Air (including the depressurisation of the accumulators), WILL the MSIV's' close on an isolation signal? EXPLAIN.
The MSIV's have a mininum and maximum closing time.
STATE the c.
minimum and maximum closing time and EXPLAIN why each of these limitations is required.
Answer 2.02
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e a.
No.
Loss of power to the DC solenoid #1 allows it to remain in the doenergized position.
Flow from the #2 AC solenoid valve will pass through the #1 DC solenoid valve and pressurine the header.
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b.
Yes. Spring pressure alone is sufficient to close the HSIV, with no
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pneumatic supply.
c.1 Minimum 3 seconds. Minimises pressure buildup in the reactor vessel due to the determination of steam flow.
- 2 Maximum 10 seconds.
Limits Off Site Dose Rates in the event of a steam line break outside the Drywell.
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Reference 1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 26; Steam Systems, pp. 23 to 25 J.
Learning Objective 828.26 3-1
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NRC Ouestion. Answer and Reference
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2.02 (Continued)
Faellity Coment
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MSIV's will close on a loss of air to the valve, with or without an isolation signal (Part b).
Should accept answer stating that the MSIV's will already be closed.
References OPM Module 26, p. 26-24
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HEC _Qusation. Answer and Referenta
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Question 2.03 The Core Spray System has initiated on Low Low Reactor Water Level.
DESCRIBE THE EFFECTU which occur on the following components as result of the initiation signal.
Include any time delays and/or equipment failures, a.
Emergency Diesels
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b.
Priority Core Spray Pumps (A and B)
c.
Core Spray Pumps C and D d.
Priority Booster Pumps (A and B)
3.
Booster Pump C and D f.
The parallel isolation valves
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Answer 2.03
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a.
The Emergency Diesels start.
b.
Priority Core Spray Pumps (A and B) start.
Core Spray Pumps C and D start in ten (10) seconds if Priority Core j
c.
Spray Pumps A and B fall to start.
d.
Priority Booster Pumps (A and 3) start.
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e.
Booster Pumps C and D start five (5) seconds after Booster Pump A or 5 falls to start.
f.
The parallel isolation valves open when pressure reduces to 285 psig.
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NRC Ouestion. Answer and Reference
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2.03 (Continued)
Reference 1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module los Core Spray System, pp. 33 to 34 3.
Learning Objective 828.10 D l
Facility Comment
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Should also accept tas actual opening setpoint for the parallel valves in the Core Spray System (Part F) which is 300f.
References Standing Orders #1
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NRC Ouestion. Answer and Rafaranga
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Question 2.04 An ELECTRICAL FAILURE has ccused an Electromatic Re*lef Valve (EMRV) to open while at 100% power.
a.
WHAT would be the FIRST indication received in the Control Room that a relief valve was open?
b.
LIST two (2) indications, available to the Centrol Room Operator, that will determine WHICH relief valve has opened.
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c.
HOW can the position of a relief valve in an intermediate position be determined?
Answer 2.04 a.
Control Room Alerms "EMRV Open" and/or "SV/EMRV Not Closed" would annunciate.
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b.1 The Valve Konitoring System (VMS) provides individual valve position indication on Panel 1F/2F.
- 2 Red and Green solenoid indicating lamps on Panel 1F/2F indicate OPEN sad CLOSED for individual valves.
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c.
Panel 15R has meters which indicate the position of the valve based on VMS amplifier outputs.
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Reference
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1.
Dyster Creek Nuclear Generating Station 2.
Operationo Plant Manual Modulo 05; pp. 11, 12, 13, 14 and 27 3.
Learning Objective 828.05 I l
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MRC Ouestion. Anawar and Rafarance 2.04 (Continued)
Zacility Coment Part a.
Operators are not required to know the actual names for the alarms they will receive.
Should accept reference to an alarm for the DGtV ' s.
Part b.
Can also get indication for which valves are open from the VMS Panel 15R.
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Part c.
The VMS indication on 1F/2F also shows the postulated posith n of the valve. However, since this is based on noise levels, the VMS cannot determine the actual position of an intermediate valve.
There is no accurate mechansism for deterniining actual valve position. Should accept answers referring to above.
References OCNGS Learning Objectives 828.05
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OPM Ndule 5, pp. 12, 14, 15
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NRC. Question, Answer and Reference
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Question 2.07
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The Reactor Feedwater Pumps may operate at greater than design flow conditions, in response to level control demand signals, resulting in pump runout conditions.
What PROTECTION IS PROVIDED to prevent the pump / motor frota being a.
damaged due to pump runout.
(Include setpoints if applicable)
b.
WHAT two (2) methods / conditions which will reset the pump runout PROTECTIVE FUNCTION 7 c.
HOW DOES THE SYSTEM RESPOND if the runout is reset while the valve controller output is reading upscale?
Answer 2.07
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a.
The flow control valve locks up at 2.67 x EE6 lbs/ hour.
b.1 The Valve Monitoring System (VMS) provides individual valve position
indication on Panel 1F/2F.
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2 Red and Green solenoid indicating lamps on Panel 1F/2F indicate OPEN and CLOSED for individual valves, c.
Panel 15R has meters which indicate the position of the valve based on VMS amplifier outputs.
Reference
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1.
Oyster Creek Nuclear Generating Station J
2.
Operations Plant Manual Modulo 05; pp.11,12,13,14 and 27 3.
Learning Objective 828.05 I s
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NRC Ougation, haver and maference
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2.07 (Continued)
Faellity Conrnent The runout protection setpoint (Part A) can vary depending on Condensate
Domineraliser differential pressure.
Should not require setpolut because it does vary.
Reference General Operating Procedure 230.0
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NRC Ougation. Answer and Raf arance
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Question 2.08 Reactor Recirculation Pump seal pressures are monitored to determine proper oparation of the seals.
a.
STATE what type of seal failure is indicated by the following conditions,' if they occur while operating at 100% power.
1)
The number 2 seal pressure is 850 psig.
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2)
The numbe'r 2 seal pressure is 150 psig.
b.
Under what two (2) conditions MUST a Recirculation Pump be shutdown
,
due to seal failure?
L Answer 2.08 a.1 Failure of the number one (1) seal.
,
2 Failure of thes number two (2) seal.
.
b.1 If the number one (l' seal temperature increases to 180 F.
2 If the nuraber two (2) seal temperature increases to 160 F.
Reference 1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 38A; Reactor Recirculation System, pp. 34 and 35
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3.
Learning Objective 828.38 H
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NRC Ouestion. hawar and Rafarance
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2.08 (Continued)
Facility Corrnant Part a.
Also, accept for #2, the fact that the #1 seal restrictive ortface could be plugged.
Part b.
Per OC Learning Objectives 828.38, the operator is not required to memorise the actual trip points based on seal failure.
Before reaching the required ternperatures the CRO will receive an alarm condition and the alarn response procedure will direct his action on tripping the pump.
Suggest consideration of answers based on this fact.
References OFM Kodule 36A, p. 38A-10 OC Learning Objectives 828.38 RAP 3024.01 E-7-3
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MRC Outation, Answer and Rafarance
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Question 2.09 The Standby Gas Treatment System has initiated on Lo Lo Reactor Water Level.
a.
At what water level does the system initiate?
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b.
LIST the four (4) additional initiating signals that will start the Standby Gas Treatment System.
(Include setpoints if applicable)
c.
How do the following system / components RESPOND when the Standby Gas Treatment System initiates?
1)
Outside air supply dampers.
2)
3)
Drywell coolers.
.
Answer 2.09 86 inches above the top of the active fuel (TAF).
a.
.
b.1 Reactor Building Rad Mr.altor High; 13 mr/hr.
2 North Wall High Vent Monitor Trips 70 mr/hr.
l 3 Operating Floor Nigh Radiation Trips 70 mr/hr.
4 High Drywell Pressures 3.5 psig.
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c.1 No automat.ic action.
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2 Isolation 3 No automatic action.
4145J 20.0
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NRC Out,*lon. Answer and Referengs
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2.09 (Continued)
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Reference 1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 42; Secondary Containment, pp. 47, 48 3.
Learning Objective 828.42 F and M Facility Comment
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Part b.
Should accept either Toch Spec or actual system setpoints for this question.
b.1 13 mr/hr (actual) or 17 mr/hr (T.S.)
b.2 70 mr/hr (actual) or 103 mr/hr (T.S.)
,
b.3 70 mr/hr (actual) or 100 mr/hr (T.S.)
,
b.4 3.0 f (actual) or 3.5 # (T.S.)
- Part c.
There are two outside air supply damper sets associated with the Rx Building supply fans.
One set is manually adjusted to set the pressure in the building.
The other set are on the discharge of each supply fan and close on a trip of the supply fan to prevent back flow through an Adle fan.
Should consider answers based on either or both sets of dampers.
Reference Standing Order f1
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NRC Ousation, Answer and Reference
Question 3.01 a.
WHICH three (3) scrams are defined as part of the MANUAL scram circuit?
I b.
During initial core loading, additional reactor protection is provided by removing the non-coincident jumpers.
WHAT effect does this have on the RPS system 7 c.
DESCRIBE the EFFECT on the RPS system if the Mode SwiEch is in STARTUP, reactor pressare is set at 700 psig and an IRM Switch is placed in range ten (10).
Answer 3.01 i
a.1 Hanual 2 Mode Switch to SHUTDOWN
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3 Non-coincident scram (Initial Fuel Loading)
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b.
Changes the RPS response to Nuclear Instrumentation from a one-cut-of-two-taken-twice logic to a non-coincident one-out-of-twenty logic for a reactor scram.
c.1 Plac.ing an IRM in range ten (10) with the Mode Switch not in RUN causes a Main Steam Lins Isolation.
L 2 With the Reactor pressure below 825 psig, the Main Steam Line.
Isolation results in a reactor scram.
,
Reference
,
1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 37; Reactor Protection System, pp. 26, 42 and 44 3.
Learning Objectives 828.37 1-A and 1-D 4145J 22.0
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NRC Ouestion, Answer and Rafaranea
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3.01 (Continued)
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Facility Cornmggt Non-coincidence jurnpers (Part a) are not covered as part of the scram circuitry because they have been installed since the initial fuel load and this scram probably will never be used again.
Suggest deleting this portion of question.
Reference OC Les.non Plan 2610.828.37
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130 Ouestion. Answer and Reference
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Question 3.02 The Oyster Creek plant is operating at 80% power when the reactor level signal to the Feedwater Level Control System is lost.
,
WHAT is EFFECT, if any, on the following components / parameters due to the loss of the level signal?
(Assume no operator action for two (2) minutes.)
a.
Actual level
.
b.
Feed flow c.
Feedwater Regulating Valves d.
Main Turbine Answer 3.02 e
a.
Level increases b.
Tsed flow increases
.
Feedwater Regulating Valves lockup (on Feed Pump Turbine runout)
c.
{
d.
Trips (on high reactor vessel level).
i Reference 1.
Oyster Creek Nuclear Generating Station
,2.
Operations Plant Manual Module 28; Feedwater Control, pp. 19, 20 3.
Learning Objectives 828.18 F-c EAC111ty_CDEtnant
Answer Key references feed pump turbines.
OC has motor driveu feed pumps..
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Reference OPM Module 17 4145J
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LRC Ouantion. Ansvar and Rafarangg
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Question 3.03 Core Spray Pipe Break Datection is provided for the Oyster Creek Core Spray System in the form of differential pressure instrumen.stion.
Unwer the following questions concerning the break detection system.
a.
What is the NORMAL DITTERENTIAL PRESSURE indicated by the system?
b.
HOW would it be possible to determine if a line break vere BETWEEN the CORE' SHROUD and the VESSEL WALL, OR in the DOWNCOMER ANNULUS?
Answer 3.03 a.
2.0 inches of water (+/- 0.5 inches)
b.
A break between the core shroud and the vessel wall would indicate a delta P of approximately 6 paid while a break in the 'downcomer would
,
indicate approximately 8-10 psid (due to the differential pressure across the stenta dryer and separator).
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Reference
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1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manual Module 55; Reactor Vessel Instrumentation System, pp. 23, 24
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Facility Conveent
.
Part b of this question caused much confusion as to actual areas in RPV being discussed. Suggest deleting Part b of this question.
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NRC Ousation. Anawar and Rafarance
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Question 3.05 l
a.
LIST the intiating signals for the Containment Spray System.
b.
LIST the equipment that automatically starts on initiation of the Containment Spray System.
(Include the time frame when equipment starts.)
Answer 3.05
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a.1 Low Low Reactor Water Level 86" above the top of active fuel TAF AND 2 High Drywell Pressure 1.85 psig.
b.1 Containment Spray Pumps A nd C start 34-46 seconds af ter auto initiation signal is received.
2 Corner Room Cooling Fans start with the spray pump start.
.
3 Emergency Service Water Pumps A and C start 48-52 seconds af ter the Containment Spray Pumps start.
.
Reference 1.
Oyster Creek Nuclear Generating Station 2.
Operations Plant Manus! Modulo 9; Containment Spray and Emergeccy Service Water Systems, pp. 24, 15 and Table 09-2
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NRC Ouestion. Answer and Reference
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3.05 (Continued)
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faellity comment Part a.
Accept actual value or Tech Spec value for drywell high pressure signal.
3# (actual) and 3.5# (T.S.).
,
Part b.
The 5% valve also opens on an initiation of the containment spray System 20 seconds after sensing 1500 gpm flow rate.
Suggest do not take credit off for discussing this valve or auto start of corner room fans.
References Standing Order #1 i
OPM Modulo 9. pp. 12, 14 and 22
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MRC Ouestion. Answer and Reference
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Question 3.08 The Oyster Creek reactor is at 100% power.
For each of the valvo conditions listed below, STATE if a (TULL SCRAM, HALF SCRAM or NO RPS ACTION) results.
a.
TSV 2 is less than 90% open.
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b.
TSV 1 and 4 are less than 90% open.
i c.
TSV 2 and 4 are less than 90% open.
d.
Emergency trip oil pressure drops to 150 psig on TCV 1 (A).
Emergency trip oil pressure drops to 150 psig on TCV 1 (A) AND e.
TCV 1 (C).
.
.
Answer 3.08 a.
NO RPS ACTION
.
b.
NO RPS ACTION c.
ONE HALF SCRAM d.
ONE RALF SCRAM
,
e.
FULL SCRAM Reference 1.
Oyster Creek Nuclear Generating Station j
2.
Operations Plant Manual Module 37; pp. 26 through 30 3.
Learning Objective 828.37 I-N 4145J 28.0
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HRfi Duantion. Answer and Peference
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3.08 (Continued)
Facilltv Coment
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Oyster C.aek has no trip funtions off of TCV position, or ETO pressure as sensed on the TCV (Parts d and e).
The ETO pressure for the generator load reject scram 14 sensed off of 4 acceleration relays in the turbine controls system. Should at,:ept "No Action" answers for these questions or answers involving ITO and the acceleration relays.
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Reference. OPM hodule 37, p. 29
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NkC Ouattlan, Answer and Reference
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Question 3.10 For the turbine trips listed below, LIST the SETPOINT, and WHAT PROTECTION la provided by the trip.
s.
Emergency Governor.
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b.
Vacuum Trip fl.
c.
Reactor High Level.
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d.
Turbine No Load With Second Stage RSCV Not Chut.
Answer 3.19 m.
110% Prevents turbine damage due to overspeed.
j b.
22" Hg Prevents overpressurising the Main Conderiser.
.
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c.
175" TAF Protects against water damage to the main turbine clades.
,
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d.
130 MW Protects the reheaters from high differential temperatures between the tube and the shell.
I Reference r
1.
Oyster Creek Wucleat Generating Station 2.
Operations Plant Manual Hoeule 50s Main Turbine Table 2 l
3.
Learning "bjective 828.50 1-2-C
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Facility Comment
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Should also accept 20% load for Part d.
Reference OPM Module 50, Table 50-2, Page 1 of 2
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MRC Ontation. Annwar and Referenes
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Question 4.01 In accordance with Standing Order 39, due to a design deficiency of the contesinment isolation relay, on the receipt of a containment isolation signal, the operator is directed to "place the Torus sample valves and the pump selector switch to the OFF position.
l EXPLAIN WHY this action is required, and what the CONSEQUENCES could be if this action were not taken, aad the containment isolation relay failed.
.
Answer 4.01 Failure of the containment isolation relay can prevent the automatfc closure of the Torus 02 analyser valves (V-38-22 and V-38-23).
Fa11vre of the valvte to go closed could result in an off site radiation release.
Reference
.
1.
Dyster Creek Nucinar Generating Station 2.
Standing Order 39 ZAcility Comtat Failure to close these valves may breach primary containment and cause a ratesse into the Secondary Containment but may not cause a release to off site as the SBGTS is also placed in service at the sAxe time.
Accept answers Giscussing a release to the secondary Containment only.
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References Standing Order #1 and #39, Page 2
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KRC_ Question. Annwar and Rafaregga
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Question 4.02 In accordance with Oyster Creek Nuc1sar Generating Station Procedure 106, there are circumstances when the Shif t Control Room Operator has the authority to shutdown the reactor.
INDICATE YES OR N0 if the Shif t Control Room Operator has the stathority to shut down the reactor for the following circumst.ances.
a.
Verified operating parameters should have initiated a' scram and no scram has occurred.
b.
In the Control Room Operators, judgement a situation exists which jeopardises or threatens to jeopardise public or plant safety.
When verified operating parameters should have initiated a safeguard c.
system and no initiation occurred.
.
.
d.
When in the Control Room Operators opinion a vic0ation of the Technical Specifications has occurred.
.
Answer 4.02 a.
Yes
!
b.
No c.
Yes
,
d.
No
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a Reference
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Oyster Creek Nuclear Generating Station Procedure 106, p. 11 i
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MRC Oggition. Laawer and Referagga
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4.02 (Continued)
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Fac111tv Connant
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Part b.
Accert yes as this is an entry condition into RPV control which
<
directs the CR0 to scram the Rs.
r Part c.
Accept yes if Tech Spec violation discussed is an LSSE.
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i Reference RPV Control Procedure EMZ-3200.01
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REC _ Question. Anayer and Reference
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l Question 4.03
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A reactor startup is in progress at Oyster Creek.
Answer the following questions in accordance with Oyster Creek Operating Procedure 201.1 Approach to Criticality.
a.
WHAT ACTION (S) is (are) required by the Reactor Operator if the Rod North Minimiser (PWM) becomes inoperable before the first 12 rods are withdrawn? WRAT NOTITICATIONS are required to be' made?
b.
WHAT LIMITATION is placed on control rod movement, when the SRM's indicate three (3) count rate doublings of the initial count rate?
Answer 4.03 a.1 Cease rod withdrawal operations.
.
.
2 Notify the Manager, Plant Operations and the Core Manager.
,
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b.
All further control rod movements must be "notched" fro'n position 06 to 48.
j Reference l
1.
Dyster Creek Nuclear Generating Statioa Procedure 201.1, pp. 6 and 8 I
laellity Cnastat
,Part a.
Answer may deal with the calendar year startup requirements as spelled cut in Tech Specs. Should accept this answer a' c s operators are not s quired to memorise procedures.
Reference: General Operating Proceduro 201.1
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NRC Ouestion. Enawar and Reference i
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Question C.04
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The plant is operating at approximately 100% power with both RBCCW pumps running, when RBCCW T ap 1-1 trips. Consequently RBCCW return temperatures begin increasing.
Per ABN-3200'.19, "RBCCW Failure Response", WHAT are three (3)
a.
methods available to reduce the heat lead oc the RBCCW system?
b.
While carrying out steps to reduce the host load, RBCCW pump 1 2 trips. Neither pump can be restarted.
WNY does ABN-3200.19 require
~
the recirculation pumps to be tripped?
(2 reasons)
Answer 4.04 a.1 Shutdown and isolate the cleanup system.
2 Reduce power
,
3 Reduce circulation flow to minimum.
,
4 Transfer the TBCCW heat exchangers to the eleculating water system.
!
b.
Seal and bearing cooling and recircuir. ting pump motor cooling are lost.
Reference 1.
ABN-3200.19, RBCCW Failure Response, pp. 4-7
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,2.
LO TCR 828.35.4, 801.01 A.7
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MAC Ouention. Answer and Reference
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4.04 (Continued)
Ins 111ty. Coment
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Part a.
Should accept anything that will act to reduce the heat load on RBCCW system or increase cooling water flow.
Operators are not
required to memorise steps or sections of the procedures. This was discussed with the Examiner.
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i Reference OC Learning Objectives 828.35 and 801.01 i
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MRC OuaAtlon, Annv.ar and Ref.erenca
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Question 4.05 Plant conditions have made it necessary to inject Standby Liquid Control into the Oyster Creek reactor vessel.
Station Operating Procedure 304, Standby Liquid Control System Operation, lists seven (7) verifications required to be made by the Control Operator to assure proper Standby Liquid Control operation.
WNAT ARE five (5) of the verifications that are required to be'made?
J Answer 4.05 a.
The PUMP ON indicating light becomes illuminated.
b.
The SQUIBB LIGHT for the selected system becomes illuminated.
An upscale reading on the pump discharge pressure is observed.
c.
,
1 d.
The FLOW ON alarm is annunciated.
.
e.
The SQUIBB VALVE OPEN alarm is annunciated.
'
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The Cleanup System inlet valveu close.
(Valves V-16-1, V-16-2 and I
V-16-4 close)
The Standby L1 quid Control Tank level indication is decreasing.
l Reference i
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j 1.
Oyster Creek Nuclear Generating Station Operating Procedure 304, pp. 10 j
2.
Learning Objective 828.46 D l
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MRf Quantion. Anawar and Rafaranea o
4.05 (Continued)
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EAcility_Comaggt i
Part F.
Change V-16-4 to V-16-14.
l Should accept power decreasing by APRM's which are also on 4F.
i
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t Referencet DWG GE 148F444 j
OFM Module 39, pp. 31 and 45 Operating Procedure 304, Step 5.3.2, p. 10
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MRC Ouestion. Answer and Enfarggga o
i Question 4.06 Operating Procedure 305, Shutdown Cooling system Operation identifies the
following as a prerequisite for operation during shutdown.
"Loop E" Recirculation Pump is running, or the
"E" Loop Discharge Valve is Closed".
a.
WRAT IS THE REASON for this prerequisite and WHAT would be the
'
CONSEQUENCES of operating with the pump not running and the discharge valve open?
Answer 4.06 a.
The Shutdown Cooling System is connected to the
"E" Recirculation loop, the reactor vessel will be "short circuited" unless the
"E" pump is running, or the
"E" loop discharge valve is closed.
!
Inadequste cooling for the core and possible core dam' age could
,
result.
t
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Reference l
,
1.
Oyster Creek Nuclear Generating Station 2.
Operating Procedure 305, shutdown Cooling System Operation, pp. 8 & 17
i Zuc.111ty coment i
Initial problem is repressurisation of RPV. Coro damage could result at a
,later time.
Reference: GE Thermo Test, Chapt. 9, pp. 117-11'
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MRC Ogsation. Answer and Reference
Question 4.07 A Control Room fire has made it necessary to evacuate Oyster Creek Control Room.
In accordance with Procedure ABN-3200.30, Control Room Evacuation:
'
a.
WHAT ARE the four (4) actions which MUST BE PERFORMED prior to leaving the Control Room.
b.
WHAT ARE four (4) of the five (5) additional actions which SHOULD BE PERFORMED prior to leaving the Control Room.
Answer 4.07
a.1 Manually scram the reactor and verify all rods inserted to or y
beyond 02.
d 2 Trip all five Recirculation Pumps.
.
3 Close the MS!V's.
4 Trip all three Reactor Feedwater Pumps.
.
b.1 Trip the main turbine.
t i
2 Check the electrical distribution system.
'
3 Confirm transfer of the house loads to Startup Transformers l
SA and St.
4 Confirm that EDG 1 and EDO 2 start and idle.
,
5 Trip all three Condensate Pumps.
4145J 40.0 l
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ca WRC.Quantion. W uer and Refertagg
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4.07 (Continued)
Reference 1.
Oyster Creek Nuclear Generating Station
,
2.
ABN-3100.30 Control Room Evacuation, pp. 3 & 4
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Zacilltv comunant i
Part b.
Should accept "Initiate B ! solation Condenter" and "Place Page
System to General".
.
Reference ABN-32CO.30, pp. 4 & 5
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MRC...Quartien, inawer and Rafarenga i
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Question 4.08 A primary system LOCA has occurred and entry into the Eriergency Procedures is required, a.
For the following conditions, STATE which Emergency Procedures would be entered.
(More than one procedure may be required for each
'
condition)
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1)
Drywell pressure 3 psig
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Drywell temperature 150 F I
3)
Reactor Pressure Vessel level 138 inches b.
LIST the three (3) Coolant Injection Sub-systems which could be utilised to inject, if the Reactor Pressure Vessel level is between 61 and 180 inches, and coolent injection is required.
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If ONLY ONE (1) of the above Coolant Injection Sub-systems can be c.
i lined up for injection, LIST the four (4) Alternate Injectior. Sub-
systems which may be utilised to provide injection.
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Answer 4.08 l
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a.1 EMG-3200-01 (RPV Control), EMG-3200-02 (Primary Containment Ccattol)
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2 EMG-32J0-02 (Primary Containment Control)
r a
3 EMG-3200-01 (RPV Control)
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4145J 42.0
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NRC Oggttion, inawer and Reference
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Answer 4.08 (Continued)
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b.1 Condensate 2 Core Spray 1
.
3 Core Spray 2 c.1 Fire Water
.
2 Core Spray Keep Full System
-
I 3 Liquid Poison Test Tank i
4 Liquid Poison Boron Tank
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Reference d
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Oyster Creek Nuclear Generating Stution
2.
Emergency Procedur1 EMG-3200.03, p. 5
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3.
Learning objective 845.04 D, 445.10 D l
l lacility comment i
Part b and c.
With level between 61 and 180 inches there is no need to go to
!
Level Restoration Procedure.
Operators may not be familiar
,
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enough with "!ajection Sub-system" and "Alternate Injection Sub-Systems" terms to relate them to the various systems called
,
out in the procedure. Should consider all systems that could i
I be listed (i.e., CRD. Core Spray, Feed and Condensate).
Discursed with Isaminer.
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References RPV Control Procedura EMG-3200.01
,
Level Restoration Procedure EMG-3200.03
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d MAC_Qas s1105 Annwar and Rafarance
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l Question 4.09 j
According to Station Procedure 2000-ABN-3200.02, "Recirculation Pump Trip", if all five recirculation pumps trip during power operation, the operator is to confirm that all pump suction and discharge valves are open, maintain water level and scram the reactor.
a.
WHY is it necessary to scram the reactor if all the recirculation pumps trip?
,
b.
WHY do Technical Specifications require at least two (2)
recirculation loop suction valves and their associated discharge l
valves to be in the full open position whenever the reactor head is I
installed on the vessel?
c.
SP 2000-ABN-3200.02 cautions that if all the recirculation pumps
'
have tripped, no recirculation pump should be restarted until the reactor has been depressurised to atmospheric pressure.
WHAT is the e
REASON for this limitation?
,
t Answer 4.09 i
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a.
Tech Specs do not allow operation with less than four recirculation
!
loops in service.
a l
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b.
The requirement assures an adequate flow path exists from the
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annular space between the pressure vessel wall and the core shroud I
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to the core region. This assures that reactor water level I
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instrumentation readings are indicative of the water level in the
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core region.
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c.
To prevent thermal over-stress in the bottom head region of the
'i a
vessel.
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H e cuantion, instar _and.Refarance
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4.09 (Continued)
Reference
'
1.
Oyster Creek Station Procedure ABN-3200.02, 3.3.1, p. 8.0 2.
Oyster Creek Technical Specificatione, Section 3.3.F, p. 3.3-3
~
3.
Oyster Creski Station Procedure 204.1, 3.1, p. 2.0 lagility co-nt
-
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Tech Specs only require a shutdown of the reactor of less than four (4) recirc loops are in service (Part a).
oc has a scram at this point because plant j
operation is outside of design limits.
Should accept answer discussing design
!
limits also.
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Reference: Tech Specs Section 3.3.F i
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J SENIOR REACTOR OPERATOR EXAM
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l HRC_QutitlpauAnawar and Rafelangts
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Question 5.03 If reactor power is increased from 60% to 70% Ly pulling rods, explain HOW and WHY the following parameters are af fected?
(Assume that no other operator action is taken.)
a.
Rectre, pump NPSH.
b.
Cere flow.
Answer 5.0:
a.
As power increases, recirc. ratio decreases, which decreases annulus temperature, which increases pump NPSM.
j b.
As power increases, voiding increases, core op increases, which decreases core flow.
"
Reference j
GE thermodynamics, Heat Transfer and Fluid Flow, Pages 7-93 thru 7-96 and 9-45 thru 9-48.
OC Lic. Op. Annual Exam. Bank, Item Code 5-7.
'
lot Lic. 'fraining Content Record 823.02, LO A.3.
j Tacility comment l
Part a.
Should also accept discussion on recirc. ratio decrease due to
Ancrease in feedwater flow which increases subcooling and NPSil.
.
Part b.
Also accept references to 2 phase flow causing core op to increase,
and decrease core flow.
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Reference OPM Module 18, Pages 4, 20-28 and CE RTFF Manual, Chapter 9, Pages 45-50.
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)[RC Ouaationn Angyer and Ref aggrgg
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Question 5.07
)
a.
DEFINE Critical Power Ratio (CPR).
b.
NOW would Critical Power change (INCREASE, DECREASE, or NO CHANGE)
for each of the following:
1.
Reactor pressure decrease.
'
2.
Inlet subcooling increase.
DESCRIBE the mechanism by which the fuel cladding could be damaged c.
if the MCPR safety limit is exceeded.
Answer 5.07 Bundle power required to produce the onset of transition boiling l
a.
divided by the actual bundle power, b.
1.
INCREASES 2.
INCREASES c.
Transition boiling causes rapid wettlug and drying of the clad surface which results in large temperature oscillations and cyclic
,
stress resulting !,n cladding perforations.
Reference i
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OC Lic. Op. Annual Exam. Bank, Item Code 5-3.
Hot Lic. Training Content F.ecord 853.09. Los P., Q. & R.
l GE Thermodynamics, Heat Transfer and Fluid Flow, Pages 9-19 & 9-66.
e Eacility Cet J
Should also accept critical power over actual bundle power for Part a.
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i Reference GE HTTT Manual, Chapt?r 9, Pages 43 and 92.
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ti1C_ Ques 11on. Anager aM.latatanga
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Question 5.09 Briefly DESCRIBE the three (3) mechanisms which can be used to assure adequate core cooling.
STATE the order of, preference for che mechanisms and EXPLAIN why they are preterred in the stated order.
Answer 5.09 1.
Core submergeace is established by maintaining RPV water level at or above the top of active fuel (TAF).
It 1s the most preferred
~
mechanism of heat removal because indication of level above TAF provides confirmation that adequate core cooling exists.
2.
Spray cooling is established (if level cannot be maintained above
TAF) by one core spray subsystem operating at or abo've design conditions.
It is less attractive than core submergence because of
,
the inability to confirm proper in-core flow distribution.
3.
Steam cocling is established (when no source of makeup is available) by maintaining a steam updraft through the core with the l
i isolation condensors ar the EMRVs.
It is the least attractive mechanism for core cooling due to the high differential temperature between the fuel and steam (limited time duration over which steam coollag can be maintained) required and the lack of instrumentation to directly confirm steam flow.
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Reference
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Not Lic. Training Content Record 845.01, L0 A.
l Emergency Operating Procedure Fundamentals, Page 6.
Facility Commtat
.
Only two methods are considered to actually assure adequate core cooling.
j Spray cooling does not assure adequate core cooling because it cannot be verified as having the proper flow distribution to assure cooling, As such, a
j OC does not have a spray coolleg procedure.
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Reference
OC Nandout 87.03, Page 5 Tech Basis for DCPGS E0Ps, Pages 1-8 to 1-11.
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HELQueltion. Answer and IR1rlanct
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Question 5.10 a.
Following a LOCA, high dryeell temperature csn affect reactor level indications.
1.
EXPLAIN HOW and WNY indicated reactor level on a Rosemount level instrument would deviate f rom actual reactor water level following an increase in drywell temperature.
STATE whether indicated level wou13 be HIGHER or LOWER than actual level.
2.
WHAT is the major concern associated with the falso indication?
r b.
During a rapid depressurisation below 500 peig, Yarvay level
]
instruments are not used to monitor reactor water level.
!
1.
EXPLAIN HOW and WHY indicated reactor level on a Yarway level instrument would deviate from actual reactor water level during a rapid depressurisation below 500 psig. STATE whether indicated level would be HIGNER or LOWER than actual level.
t 2.
WHAT is the major concern associated with the f alse
indication?
i
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Answer 5.10
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a.
1.
As the reference leg water temperature increases, the
.
reference leg water density decreases causing the sensed
!
differential pressure to decrease.
The lower differential pressure registers as an increased vessel level.
Indicated
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level is HIGHEd than actual.
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Automatic Actioni (initiated on Lo or Lo-Lo reector level)
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could be delayed or prevented.
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b.
1.
We*,er in the heated reference legs will flash below 500 psig
,
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causing water in the reference leg to decrease causing the
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sensed differential pressure to decrease. The lower
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differential pressure registers as sa increased vessel i
level.
Indicated level is HIGHER than actual.
1 J
2.
Accurate water level indication is necessary to ensure
)
adequate core cooling is maintained.
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Reference j
OC Operations Plant Manual, Modt'le 55, Pages 11, 21 & Table 55-1.
Synptom Based Emergency Operating Procedure Basis Review, Page 7.
Hot Lic. Training Content Record 815.04, LO A.
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KQues. tion, hawar and Reference
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5.10 Continued
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Faellity C mment
l Part b. refers to Yarway level instruments.
Actual instruments are Rosemounts
but are referred to as Yarways for human factor reasons.
Should accept a disetssion based on Yarway or Rosemount.
<
Reference
>
i OPM Module 55, Page 11 and Yable 55-1.
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I HRCJ1usAtlen. Amarat And_Refstanca
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Question 6.01 The lead Control Room Operator has just placed the Liquid Poison System
.
keylock switch on Panel 4F in the "System 1" position.
,
a.
WHAT two (2) actions occur as a direct result of placing the switch in this position?
(within the Standby Liquid Control System).
b.
WHAT are four (4) indications available on Panel 4F that can be used to verify that Liquid Poison is injecting?
Answer 6. 0 '.
l a.
1.
Liquid Poison Pump 1 starts.
2.
Squib Valve 1 Fires.
b.
1.
PUMP light is on.
2.
SQUIBS light is on.
,
3.
Pump discharge pressure.
l 4.
Tank level decreasing.
-
5.
Poison inlet valve OPEN light is on.
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Reference j
OPM-Module 46 CLO. Pages 14,16, Fig. 46-4.
!
LO TCR 828.46 E.F,0.
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TAcility cemment I
l Add "power decreasing" as an answer to part b. because the Nuclear i
Instrumentation System reads out on Panel 4F.
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i Reference f
None.
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MRC Ouantion, Lnawar anA nafarangs
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Question 4.02 The Oyster Creek reactor is at 100% power.
a.
A turbine trip occurs.
HOW will each of the following sets of valves respond (OPEN, CLOSE or RCAAIN ,IS)?
1.
Turbine Stop Valves.
2.
Turbine Control Valves.
3.
Turbine Rypass Valves.
b.
For each of the valve conditions listed below, STATE if a (rttLL SCRAM, HALF SCRAM or NO RPS. ACTION) results with the plant at 100%
power.
,,
1.
TSV 1 is less than 90% open.
2.
TSV 2 and 4 are less than 90% open.
3.
Emergency trip oil pressure drops to 150 psig on TCV 1 (A)
and TCV 1 (C).
Answir 6.02 a.
1.
TSV's close.
2.
TCV's close.
3.
Turbine bypass valves open.
- b.
1.
NO RPS ACTION.
2.
ONE HALF SCRAM.
3.
FULL SCRAM.
Reference
.
Oyster Creek Nuclear Generating Station.
Operations Plant Manual Module 5), Fages 26 through 30.
OPM-Module 51: Turbine Control System, Pages 9,20,35,41.
LO TCR 828.51 D,F and 824.37 I-N.
FAC111Lyl9mtAt ETO is sensed from 4 acceleration relays, not from the TCV (Part b.3).
Should accept any discussion of ETO with reference to the seceleration relays, or "No Action".
Reference
.
OPM Module 37. Page 29, 2285C 7.0
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NRC Ouestien d awar and Refertngt
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Question 6.03 The reactor is operating at 100% and a full flow surveillance on the Core Spray System is in progress. The CS system is recirculating water back to the Torus when a CS initiation signal is received.
a.
WHAT is/are the initiation signal (s)? Setpoints required.
b.
HOW does the test-to-return valve respond to the initiation signal?
Initially, DOES the CS discharge valve automatically open? WHY or c.
NKt NOT7 Answer 6.03
-
Low Low vessel water level of 86 inches above TAF high D/W pressure a.
of 3.5 psig, b.
The valve will automatically close.
c.
No.
The Valve will temain closed as long as reactor pressurr is 285 ptig.
,
Reference l
OPM-Module 10-Core Sprav System, Pages 19-23.
.
LO TCR 828.10 C,D.
,
Facility Comment Should accept either Actual Setpoints or Tech Spec Setpoints for Part a.
.
Low Low Level 90" TAF (Actual) 86" TAF (T.S.).
.
High Dw. Pres. 3# (Actual) 3.5# (T.S.).
Fisfurence i
Standing Order fl.
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MAC Question. Angygr and Rafaranea
Question 6.04 Reactor Recirculation Pump seal pressures are monitored to determine proper operation of the seals.
,
STATE what type of seal failure, if any, is indicated by each of a.
the following conditions with the plant operating at 100% power.
1.
The number 2 seal pressure is 850 psig.
2.
The number 2 seal pressure is 150 psig.
b.
Under WHAT two (2) conditions MUST a Recirculation Pump be shutdown J
due to seal failure?
Answer 6.04
'
a.
1.
Failure of the number one (1) seal.
2.
Failure of the number two (2) seal.
b.
1.
If the number one (1) seal temperature increases to 180*F.
2.
If the number two (2) seal temperature increases to 160*F.
)
Reference
'
Oys'at Creek Nuclear Generating Station.
Operations Plant Manual Module 38A Reactor Recirculation System, Pages 34 and
.
35.
Learning Objective 828.38 H.
Facility Sommerl
_
Also accept plugging of #1 seal restrictive orifice as an answer for Part a.2.
.
Reference
OPM Module 38A, Pages 9 and 28.
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MCR Ouestion. Answer and Raf aranea
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Question 6.05 An automatic initiation signal for the Isolation Condenser System has been received and the condensers are in operation, when a steam line break occurs BETWEEN the pressure vessel and the differential pressure detectors for the
"A" condenser.
a.
MILL the
"A" Isolation Condenser isolate? EXPLAIN.
b.
On an isolation signal, WILL a LOSS of 480 volt AC power PREVENT'
the ISOLATION of the condenser? EXPLAIN.
c.
WILL a LOSS of Instrument and Service Air PREVENT automatic makeup to the Isolation Condenser? EXPLAIN.
,
Answer 6.05 a.
Yes. The differential pressure detectors respond to a differential pressure (with steam flow passing in either direction).
i b.
No.
Each steam supply and condensate return line has two (2) MOV l
1 solation valves in series. One valve is 480 volts AC and the I
other is 125 voitt DC.
On loss of 480 volt AC power the DC valves l
will still close.
!
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c.
No.
An air accumulator provides air to operate the makeup valves at least sin (6) times with the complete loss of Instrument and
.
Service Air.
,
Reference Oyster Creek Nuclear Generating Station.
.
Operations Plant Manual Module 23 Isolation Condenser System, Pages 9, 10 and 13.
Learning Objective 828.23 I.
Egellity comment In Part c., the question refers to automatic makeup to the Isolation Condenser. OC does not have an automatic makeup to the IC's.
Makeup is a remotely controlled function.
Loss of air does not prevent remote operation of the makeup valve for up to six lines.
Reference
,
OPM Module 23, Page 23-17.
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o-NRC Outit10n. Answer and Referengt
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Question 7.04 The plant is shutdown for refueling.
The reactor head is removed and the reactor cavity is flooded above the main steam nossles.
In preparation for fuel offload and maintenance on the recirculation pumps, your shift has been instructed to secure and isolate all five recirculation pumpa in accordance with Procedure 301, "Nuclear Steam Supply Systen".
a.
A temporary change to Procedure 301 has been written to allow isolation of all five recirculation loops simultaneously.
EXPLAIN why this can be done without violating a safety limit.
b.
Procedure 301 allows operation of the recirculation pumps under two conditions when fuel is removed from the core.
STATE the two (2)
conditions.
Answer 7.04 a.
The safety limit does not apply since the head is removed and the cavity is flooded above the main steam nossles.
b.
1.
If it is necessary to initiate Standby Liquid Control.
2.
All LPRMs, IRMs and SRMs in the vessel are surrounded on all sides by fuel assemblies and/or guide blades.
Refer 1nce j
Procedure 301, "Nuclear Steam Supply", Page 22.
LO TCR 828.38 G.
Facility Comment Answer requires operator to have memorized a section of the procedure.
This is not required per our Learning Objectives.
Should allow leeway on answer in light of '21s.
Coment discussed with examiner.
Reference OC Learning Objectives 828.38.
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NRC Outstlen. Ansvar and Rafaranca
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Question 7.06 A containment isolation signal has been received.
Due to a design deficiency in the torus oxygen sample line isolation logic, Standing Order 39 requires that the Torus sample valves and the pump selector switch are placed in the OFF position on the receipt of a containment isolation signal.
a.
EXPLAIN WHY this action'is required and WHAT the CONSEQUENCES could be if this action were not taken and the containment isolation relay failed, b.
An emergency condition exists which necessitates bypassing of the isolation signal.
1.
WHO (by title) must approve bypassing of the isolation signal?
2.
TRUE or FALSE 7 Procedure 312.1, "Bypassing Isolation Interlocks During Emergency Conditions', controls the installation and removal of any jumpers required to bypasa the isolation signal.
Answer 7.06 Failure of the containment isolation relay can prevent the a.
automatic closure of the Torus 02 analyser valves (V-38-22 snd
.
V-38-23).
Failure of the valves to go closed could result in an off site radiatton release.
b.
1.
The Emergency Director or the Group Shift Supervisor.
2.
FALSE (prior to removal control must be transferred to
.
Procedure 108, "Egulpment Control").
j Reference OCNGS Standing Order 39, License Event Report (LER)87-040.
i Procedure 312.1, "Bypassing Isolation Interlocks During Emergency Conditions",
Page 3.
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I LO TCR 830.05 5.
Facility comment
.
j Part b.1.
This requires memorisation knowledge of Procedure 312.1 and is not required by our Learning Objectives.
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s NRC Ogallion, hawar and Referenga.
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Question 7.06 Continued Facility Comment Continued Part b.2.
Sign off pagse for jumpers 4031 control the removal of the jumpers in that there is a sign off block verifying that control has been transferred to Procedure 108. Therefore, should consider True as a correct answer. Also this requires intimate knowledge of the procedure which is not required by our Learning Objectives.
i Comment discussed with examiner.
Reference Operating Procedure 312.1, Page E9-2.
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OC Learning Objectives 845.02.
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Question 7.07 The piant is operating at approulmately 100% power with both RBCCW pumps running, when RBCCW pump 1-1 trips. Consequently RBCCW return temperatures begin increasing.
a.
Per ABN-3200.19, "RBCCW Failure Response", WHAT are three (3)
methods available to reduce the heat load on the RBCCW system?
l b.
While carrying out steps to reduce the heat load, RBCCW pump 1 2 trips. Neither pump can be restarted.
WHY does ABN-3200.19 require the recirculation pumps to be tripped?
(2 reasons)
Answer 7.07
-
a.
1.
Shutdown and isolate the cleanup system.
2.
Reduce power.
3.
Reduce recirculation flow to minimum.
4.
Transfer the TBCCW heat exchangers to the circulating water
system.
b.
Seal and bearing cooling and recirculating pump motor cooling are lost.
Reference ABN-3200.19, RBCCW Failure Response, Pages 4-7.
LO TCR 828.35 4, 801.01 A.7
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Facili_tv_ Comment Should accept anything that will act to reduce the heat load on the RBCCW
system or increase cooling water flow. Operators are not required to memorise i
steps or sections of the procedures.
This was discussed with the examiner.
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Reference OC Learning Objectives 828.35 and 801.01.
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RC_Quantion, Aneve r and nef e rerla
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Qucation 7.08 For each condition / situation, STATE which Emergency Operating Procedures, if any, would be entered.
If none, state NONE.
a.
Drywell pressure of 2.3 poig.
b.
Rs. Building ventilation exhaust radiation of 15 mr/hr.
c.
Bulk Drywell temperature of 153 degrees T.
d.
Torus water level of 162 inches.
e.
Load Reject has occurreo and power is 40%.
f.
RPV water level is 125 leches above TAF.
Answer 7.08 a.
None.
b.
EMG-3200.11, Secondary Containment Control.
c.
EMG-3200.02, Primary Containment Control.
d.
EMG-3200.02, Primary Containment Control.
i e.
EMG-3200.01, RPV Control.
f.
EMG-3200.01, RPV Control.
Reference Emergency Operating Procedures EMG-3200.01, 3200.02, 3200.11, Page 3.
LO TCR 845.03 B, 8450.11 B, 845.06 B.
Facility comment
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Part 6 Should be changed to "None" in the answer key as 35% is within the capacity of our bypass valves to handle for a finite time period (40%) and the load reject scram is bypassed below this point -
therefore, no scram is required. Discussed with examiner.
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Reference RPV Control Procedure ENG-3200.01.
OPM Modulo 37, Pages 29-31.
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NRC Outation. Answer and Referents
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Question 7.09 Emergency Procedure EMG-3200.09, "Lovel/ Power Control" requires that reactor water level be lowered by securing injection systems it Power is above 2%, AND
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Torus water temperature is above the Boron injection temperature,
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AND Either an EMRV is open or drywell pressure is above 3.0 psig.
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a.
WHICH of the injection systems WOULD NOT be secured in order to
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lower water level?
b.
Reac*,or water level will continue to be lowered until one of three (3) conditions is met. WHAT ARE the three (3) conditions?
I Answer 7.09 s.
1.
Boron injection.
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2.
CRD.
b.
1.
Reactor power drops below 2%.
2.
Reactor water level reaches 0 inches.
3.
All EMRV's remain closed and drywell pressure remains below
3.0 psig.
i Reference
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EMG-3200.09, Page 5.
LO TCR 845.19 B.
I FAC111tLCQEntat
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i To ansvar this question twquires that the procedure be memorised which is not j
required by our Learning Objectives.
Should be vary broad in viewing the i
answers received on the questions, also consider conditional statements.
I Discussed with examiner.
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OC Learning Objectives $45.19.
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se MRC Ouantion. Annwar and Rafarents
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Question 7.10 According to Procedures IMP-1300.02, "Direction of Emergency Response" and
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IMP-1300.03, "Emergency Notification":
a.
(TkUE/ FALSE) The Emerge;.Jy Director (ED) is responsible for classifying the event unless overruled by the Emergency Support Director (ISD).
b.
(TRUE/ FALSE) Once an event has been classified and the NRC has been i
notified of its classification, NRC permission is required to
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change the emergency's classification.
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c.
(TRUE/ FALSE) Even with the ESD function activated, the ID is still
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responsible for epproving and directing information releases to the media.
d.
(TRUE/ FALSE) The ED shal' fo13cw the NRC's advice to devitte from a
established operating procedures during attempts to control the
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emergency, i
e.
(TRUE/ FALSE) In the event that the NRC and the ED/ESD have differing recommendations for courses of protective action, the Plant Manager will resolve the conflict.
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l Answer 7.10 i
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a.
True
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b.
Falso c.
Falso
d.
Falso
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e.
Falso
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Refe,ence
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IMP-1300.02, Pages 2-5.
j IMP-1300.03, Page 5.
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IaelliLY_Cnement i
The answer to Part a. is in reality falso because the only time the ESD can i
override the ED is in situations where the ESU wants to classify an event at a higher level than the ID.
The ESD cannot prevent the ED from escaleting an
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, emergency condition.
Reference l
OC Procedure IMP-1300.02, Page 4.
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Question 8.02
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Procedure 126. Procedure for Notification of Station Events, defines sis categories of events and specifies appropriate notification requirements.
a.
DEFINE Category I events.
b.
HOW will you determine what notifications must be made for Category I events?
c.
HOW LONG do you have to notify the NRC after a Category !! event.
occurs?
Answer 8.02
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a.
This category lecludes all events which result in declaration of an Unusual Event. Alert, site Area Emergency, or a General Emergency.
b.
Procedure 126 directs you to appropriate EPIPs which specify the notifications that must be made.
c.
One hour.
Reference OC Procedure 126, Procedure for Notification of Events, Page 4.
OC Lic. Op. Annus! Imam. Bank, Item Code 4-56.
Not License Training Content Record 8?0.05, LO: XX.
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Facility Comment i
Should also accept the general definition of "all events classified by the
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EPIP's".
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i Reference None.
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MRC.Quantion. Anawar and Rafaranea
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Question 8.04 a.
WHAT is required to maintain a Senior Reactor Operater (SRO)
license in an ACTIVE status in accordance with 10 CFR Part 557 b.
If a person's SRO license is in an INACTIVE status, what is required before he or she can resume duties that require an active SRO license?
c.
An unlicensed operator wishes to operate the controls to insert a rod during a rod pattern exchange operation under the direction of the lead Control Room Operator.
Is he allowed to do this? WHY or WNY NOT7
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Answer 8.04 l
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The license holder must perform the functions of an SRO on a
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minimum of 7 eight hour shifts or 5 twelve hour shifts per calendar quarter.
b.
The license holder must perform a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift
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functions under the direction of an SRO. These functions must l
include a tour of the plant and all shift turnover procedures.
(Yes or No) He may perform this function only if it is part of a c.
licensed operator training program.
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Reference
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10 CFR Part 55, Sections 13(a) (2), 53(e) and 53(f).
i raellity comment
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Should not weigh the time requirements for SRO license requirements as heavily
j ss they are. Should have most credit based on functions necessary to maintain a license.
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Reference i
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MAC_QRg1119A&wer and Raingue
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Question 8.09
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I A startup is in progress with the mode switch in STARTUP.
The
'A' Offgas
Process Radiation Monitor has been inoperable for one week.
The replacement parts are on order.
On your shift the
'B'
Offgas Process Radiation Monitor failed to meet the surveillance test acceptance criteria.
Repairs and i
recalibration is expected to take until the end of the shift.
Can you, as the Group Shift Superviscr, direct the Startup to continue?
If it can continue, can the mode switch be taken to RUN when plant conditions permit? EXPLAIN.
NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECITICATIONS TO ANSWER THIS QUESTION.
TULLY REFERENCE ALL APPLICABLE SECTIONS OF THE TECHNICAL SPECITICATIONS.
Answer 8.09 l
Yes the startup can continue and the mode switch can be taken te RUN.
TS Table 3.1.1 requires one trip system operas?h in STARTUP and 2 channels within the trip system operable. Note 11 allows removal of one channel within the trip system for maintenance. With no channels operable note jj allows i
continued operations for 72 hrs. provided that the stack noble gas monitor is
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ope r able.
Reference Oyster Creek Technica4 Specifications Table 3.1.1, Hot Lic. Training Content Record 850.90 LO J.1 and J.2.
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t facility coment
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Note 11 of Table 3.1.1 is not applicable, but overridden by Note jj.
Credit j
should not be taken oft for not referring to Note 11.
Reference Tech. Specs. Table 3.1.1.
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